Experimental study of abrupt discharge of water at supercritical conditions

Future SuperCritical Water-cooled nuclear Reactors (SCWRs) will operate at a coolant pressure close to 25 MPa and at outlet temperatures ranging from 500 degrees C to 625 degrees C, i.e., above the critical pressure and temperature of the water (22.06 MPa and 373.95 degrees C, respectively). Coolant pressures higher than critical values will be used to avoid boiling and eventual critical heat flux that may occur. In addition, the outlet flow enthalpy in future supercritical water-cooled nuclear reactors will be much higher than those of actual ones, which can increase overall nuclear plant efficiencies of up to 48%. However, under such flow conditions, thermal-hydraulic behaviors of supercritical water are not fully known, i.e., pressure drop, the deterioration of forced convection heat transfer, critical (choked) flow, blow-down flow rate, etc. In particular, the knowledge of critical discharge of supercritical fluids is mandatory to perform nuclear-reactor safety analyses and to design key mechanical components. Nevertheless, existing choked-flow data have been collected from experiments at atmospheric discharge pressure conditions, but in most cases using working fluids different than water. Therefore, a supercritical water facility has been built at the Ecole Polytechnique de Montreal. In this paper, a new database containing 524 data points is obtained using this facility and compared with available information from the open literature. Crown Copyright (C) 2014 Published by Elsevier Inc. All rights reserved.